The present invention relates to novel austenitic stainless steel, especially, to novel structural austenitic stainless steel preferable for use in a radiation irradiated environment, such as nuclear reactor cores and nuclear fusion reactors, and the usage of the same.
Austenitic stainless steel, especially the stainless steel having a high chromium-nickel composition, has been used as a material for structural members in nuclear reactors, because the stainless steel has a corrosion resistance in a corrosive environment in addition to having preferable properties as a structural material. However, a member made of austenitic steel in light water reactor cores becomes increasingly sensitive to Intergranular Stress Corrosion Cracking (IGSCC) with long time irradiation in use. For instance, a stainless steel having a solid solution state, obtained by solution heat treatment, has strong resistance to IGSCC outside the reactor core where no radiation damage is caused. However, the same material loses resistance to the IGSCC when it is irradiated with high level radiation, especially radiation at more than 0.5.times.10.sup.21 n/cm.sup.2 in neutron dose, inside the reactor core. The IGSCC phenomenon is called Irradiation Accelerated Stress Corrosion Cracking (IASCC), has currently been deemed as a problem relating to aged nuclear reactors. As for the metallurgical mechanism of the IASCC phenomenon, two mechanisms of this phenomenon have been well known, including (1) atomic diffusion (irradiation induced diffusion), accompanied by a migration of voids which are generated by irradiation, causes a reduction in the concentration of chromium, a corrosion resistant element, in the vicinity of grain boundaries, and (2) impurity elements, such as P, S, si, and others, are segregated at grain boundaries, which reduces corrosion resistance at the grain boundaries.
As for a method for resolving the above described problems, JP-A-63-303038 (1988) discloses a method where the amount of composite elements of austenitic stainless steel, such as N, P, Si, S, C, Mn, Cr, and Ni, are adjusted, and traces of Ti and Nb are added to the austenitic stainless steel.
On the other hand, as for a method for preventing intergranular stress corrosion cracking, single crystallizing methods for eliminating grain boundaries, which form a network and are a source of the cracking have been proposed. As for the single crystals, a single crystal steel of austenitic (.gamma.) single phase having a crystalline structure of FCC structure, a single crystal steel of austenitic phase matrix including a small amount of ferritic (.delta.) phase having a crystalline structure of BCC structure, and a single crystal steel of a so-called two-phase stainless steel, wherein the (.gamma.) phase is dispersed in a single crystal of .alpha. phase, are disclosed in JP-A-3-264651 (1991) and JP-A-62-180038 (1987).
However, the invention disclosed in JP-A-63-303038 (1988), wherein intergranular stress corrosion cracking is prevented by adjustment of the composition, can not prevent substantially all of the above described stress corrosion cracking generated by irradiation acceleration, because the material is polycrystalline stainless steel and its structure contains grain boundaries.
Furthermore, the proposal disclosed in JP-A-3-264651 (1991) relates to a single crystal steel of .gamma. single phase including steel with Ti, Nb added, or a .gamma. phase single crystal steel containing a small amount of .delta. phase, both of which steels are strengthened by a solid solution of alloy elements and have a lower yield strength than that of commercially available stainless steels, SUS 304 steel and SUS 316 steel. The proposal disclosed in JP-A-62-180038 (1987) relates to a single crystal, of which the parent phase is composed of a .alpha. phase having a BCC structure. The .alpha. phase has been a source of concern as it is conceived to more readily cause irradiation embrittlement by irradiation damage than the .gamma. phase.